• Nuclear Engineering and Technology
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• 제21권1호
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• pp.48-55
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• 1989

본 논고에서는 국내의 PWR 및 CANDU 사용후 핵연료로부터 발생하는 붕괴열의 장기적인 거동을 보다 손쉽게 분석하기 위하여 붕귀열을 추정할 수 있는 간단한 근사식을 도출하였다. 근사식의 장기적인 붕괴열 추정에서 ORIGEN 2코드 결과와의 차이를 줄이고 중요한 변수 조건하에서도 붕괴열을 추정할 수 있도록 하기 위하여 민감도 분석을 수행하였다. 그 결과로서 얻어진 근사식은 사용후 핵연료의 이력자료중 중요변수인 연도를 포함함으로써 3~500년정도의 냉각시간 범위내에서는 임의의 연소도를 가진 사용후 핵연료의 붕괴열이라도 추정할 수 있게 되었다. 그리고 대표적으로 30, 37 및 40 GWD/MTU등의 연소도를 갖는 사용후 핵연료의 붕괴열 추정에 있어서는 1년부터 $10^{5}$ 년까지의 냉각시간에 따라 ORIGEN2 ,코드의 결과와 $\pm$10%이내의 차이를 보이고 있어 사용후 핵연료 관리를 위한 관련시설의 열적설계 및 평가 등과 같은 공학적 목적에 유용하게 사용될 수 있을 것이다.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.79-88
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• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.674-679
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• 1995

The hold-down spring forces of Kori nuclear fuels were measured for seven fuel assemblies having 1 to 4 cycles of irradiation histories in the Kori Unit-1 and -2 reactor. The fuel assemblies examined had burnup from 17 to 38 GWD/MTU and the examination was conducted in KAERI PIEF spent fuel storage pool with the newly developed underwater hold-down suing force measuring device. The measurement was made within the elastic deformation ranges and the trends of hold-down spring force relaxation behavour were examined.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2007년도 추계학술대회 논문집
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• pp.439-440
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• 2007

• Bulletin of the Korean Chemical Society
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• 제24권11호
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• pp.1686-1688
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• 2003

• 한국표면공학회지
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• 제56권1호
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• pp.84-93
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• 2023

The spent nuclear fuel is burned during the planned cycle in the plant and then generates elements such as actinide series, fission products, and plutonium with a long half-life. An 'interim storage' step is needed to manage the high radioactivity and heat emitted by nuclides until permanent-disposal. In the case of Korea, there is no space to dispose of high-level radioactive waste after use, so there is a need for a period of time using interim storage. Therefore, the intensity of neutrons and gamma-ray must be determined to ensure the integrity of spent nuclear fuel during interim storage. In particular, the most important thing in spent nuclear fuel is burnup evaluation, estimation of the source term of neutrons and gamma-ray is regarded as a reference measurement of the burnup evaluation. In this study, an analysis of spent nuclear fuel was conducted by setting up a virtual fuel burnup case based on CE16×16 fuel to check the total amount and spectrum of neutron, gamma radiation produced. The correlation between BU (burnup), IE (enrichment), and CT (cooling time) will be identified through spent nuclear fuel burnup calculation. In addition, the composition of nuclide inventory, actinide and fission products can be identified.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4499-4513
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• 2022

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel-specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM^TM). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ~50 years of dry storage.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2000년도 추계학술대회 논문집
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• pp.441-444
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• 2000

The recycling process for reuse of uranium in the spent fuels consists various unit processes and the decladding process to extract the spent fuel pellet from the zirconium-based cladding is the beginning process of the recycling. There are two methods - mechanical and chemical - in the decladding process. In this paper, the mechanical decladding device by using a motor as a driving part and a press pin to separate the pellets from tube has been developed. This device was automated and modularized to make the remote operation and maintenance easy.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2003년도 춘계학술대회 논문집
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• pp.1438-1441
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• 2003

The project for "Development of Advanced Spent Fuel Management Technology" has a plan of a demonstration for the Advanced Management Process in the hot cell of IMEF. The Advanced Management Process are being developed for efficient and safe management of spent fuels. For the demonstration, several devices which are used to safely transport and handle nuclear materials without scattering have been derived by analyzing the Advanced Management Process, object nuclear material and modules of process equipment and performing graphical simulation of transportation/handling by computers. For verification, powder transportation vessel and handling device have been designed and manufactured. And several tests such as transporting, grappling, rotating the vessel have been performed. Also, the design requirements of transportation/handling equipment have been analyzed based on test results and process studies. The developed design requirements in this research will be used as the design data for the Advanced Management Process.

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1307-1314
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• 2009

The isotope ratio of $^{134}Cs/^{137}Cs$ in a spent PWR fuel sample was obtained with a newly developed gamma/neutron combined measuring system at KAERI. Burnup and Pu/U ratio of the spent fuel sample were determined by using the measured isotope ratio and the burnup-isotope ratio correlation equations calculated from the ORIGEN-ARP computer code. The results were compared and evaluated with the chemically determined burnup and Pu/U ratio. As a result of the comparative evaluation, the nondestructively determined burnup and Pu/U ratio values showed a good agreement with the chemically obtained results to within a 4.5% and 0.8% difference, respectively.